ML20044H040
| ML20044H040 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 05/26/1993 |
| From: | Selin I, The Chairman NRC COMMISSION (OCM) |
| To: | Daley M NEW ENGLAND COALITION ON NUCLEAR POLLUTION |
| Shared Package | |
| ML20044H041 | List: |
| References | |
| NUDOCS 9306070270 | |
| Download: ML20044H040 (16) | |
Text
'
- x Enclosure e
/
o UNITED STATES g
NUCLEAR REGULATORY COMMISSION g
y WASH:NGTON, D. C. 20565 o.,
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May 26, 1993
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- CHAIRMAN Mr. Michael J. Daley New England Coalition on Nucitar Pollution, Inc.
Box 545 Brattleboro, Vermont 05301
Dear Mr. Daley:
On behalf of the Commission, I am responding to your letter of February 9, 1993, in which you expressed disappointment in the NRC Region I Administrator's answer to a letter you sent him on December 16, 1992.
I am also acknowledging your letters of April 8 and 11,1993, in which you expressed concern about the emergency diesel generators at the Vermont Yankee Nuclear Power Station. These two letters are being handled separately as a petition submitted under 10 CFR 2.206 and have been referred to the Director of the Office of Nuclear Reactor Regulation. The Commission and the NRC staff, including our Region I office, recognize the seriousness of the concerns you have raised, respect your views on equipment performance and other reactor safety issues at the Vermont Yankee Nuclear Power Station, and share your interest in the continued safe operation of the plant.
The Region I Administrator's letter of January 25, 1993, did not deal lightly with the issues you raised.
The staff evaluated the information in your letter, but believed these issues had been adequately treated in the past. A summary of the approximately 40 events that you cited in your December letter, including the NRC's independent evaluation of both root causes and corrective actions, is provided in Enclosure 1.
The Commission too is concerned about patterns of poor performance and declining performance trends.
For this reason, the NRC inspection and assessment processes are performance oriented. The adequacy of a licensee's support of a plant (material, resource, and otherwise) is assessed on a continuing basis through independent inspections, the semi-annual senior NRC management review process, and our systematic assessment of licensee performance (SALP) program. The type, degree, and schedule of inspections are reviewed at least semi-annually for each facility. Adjustments are made consistent with the licensee's performance as observed during inspections.
The most recently completed SALP evaluation at Vermont Yankee covered the period from March 1991 to August 1992 and reached conclusions about functional areas in which performance was rated as superior, where it was observed to be good and improving, and areas that showed declining performance.
The SALP Board concluded that Vermont Yankee conducted its activities in a safe manner.
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The NRC's SALP evaluations, including the most recently issued report for Vermont Yankes, are an integrated assessment of inspection findings, enforcement actions, plant performance, and reportable events, such as those cited in your December letter. The SALP Board looks for common root causes indicative of potential problems, either by functional area (e.g., maintenance and surveillance) or cause code. Such causal analyses of reportable events are integral to the SALP process. A summary breakdown of all Vermont Yankee Licensee Event Reports (LERs) spanning the 6-year period of 1987 through 1992 is listed in Enclosure 2.
This analysis encompassed the last four SALP reports issued by the NRC.
In fact, LERs were typically evaluated individually during NRC inspections as well as collectively in the SALP process.
During the 6-year period of 1987 through 1992, there were 145 inspection reports issued involving more than 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of on-site independent inspection. We believe the results summarized, both qualitatively in our SALPs and quantitatively in Enclosure 2, support the conclusions stated in our earlier letter to you relative to equipment degradation.
The NRC will continue to monitor Vermont Yankee's performance.
In the near term, inspection initiatives have been undertaken specifically to evaluate the Vermont Yankee Nuclear Power Corporation's resolution of the fire barrier issues, emergency diesel generator reliability problems, and control rod drive surveillance test anomalies.
Further, the staff has scheduled a Service Water System Operational Performance Inspection for later this year to confirm system performance and plans to conduct an Operational Safety Team Inspection in June 1993 to further evaluate performance across several functional areas in an integrated fashion.
We want to assure you that any new information regarding Vermont Yankee's performance developed through our inspection process or from any other source will be evaluated promptly and that effective corrective actions will be taken as warranted by circumstances.
Sincerely,
/
Ivan Selin
Enclosures:
1.
Summary of Events 2.
Vermont Yankee LERs, 1987-1992 i
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SUMMARY
OF EVENTS
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All licensee event reports (LERs) and other issues cited in the December 16, 1992, letter from the New England Coalition on Nuclear Pollution were (NECNP) previously reviewed and inspected by the NRC.
No new issues are evident and the events encompass a wide range of areas.
Generally, the. licensee's corrective actions were independently found by the NRC to be adequate.
Issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Inadvertent emergency NRC Inspection Root cause attributed to an core cooling system Report (IR) 89 02, informal restoration procedue-(ECCS) actuation while LER 8915 The NRC determined that the repowering the "B" licensee's corrective actions were ECCS logic system.
adequate (personnel training, procedure development).
Inadvertent core spray IR 89-07, LER
._ Root cause attributed to an (CS) and residual heat 89 16 inadequate test procedure. The removal (RHR) pump NRC determined that.the licensee's start due to inadequate corrective actions wem adequate procedure.
(design review, procedum modification ).
High. pressure coolant IR 91-11, IR Root cause attributed to a injection (HPCI) 9129, LER 91-07 component failure. The NRC inoperable (INOP) due determined that the licensee's to flow controller corrective actions were adequate setpoint drift.
(flow controller replaced, additional monitoring of controller setpoints).
HPCI INOP due to IR 92-06, IR Root cause attributed to component.
degraded battery bus 92-04, LER 92-04 failure. The NRC determined that -
voltage.
the licensee's cornetive actions were adequate (alarm response sheet avision, component replacement, engineering review, and personnel training). The NRC issued a Notice of Violation to the.
licensee for falling to make the required 10 CFR 50.72 report within 41.ours.
A, l
2 1
issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Reactor core isolation IR 89-09, IR Root cause attributed to component cooling (RCIC) INOP 89-12, IR 89 22, failure due to excessive cycling over due to a MOV failure.
and LER 89-14 a short period of time. The NRC determined that the licensee's corrective actions were adequate (component replacement and failure analysis, procedural enhancements).
RCIC INOP due to flow IR 92-12 and LER Root cause attributed to component controller drift.
92-15 failure. The NRC determined that the licensee's corrective actions were adequate (controller calibration and increased setpoint monitoring).
Inadvertent reactor IR 90-03, IR 9122 Root cause attributed to personnel scram de to a short and LER 90 09 error. The NRC determined that circuit on the vital AC the licensee's corrective actions bus which was caused were adequate (personnel training).
by personnel error.
Overloaded power IR 8814, IR 88-19 Root cause attributed to a design supply in fire protection and LER 8812 deficiency. The NRC determined control panels.
that the corrective actions were adequate. The NRC issued a Notice of Violation to the licensee for inadequate design control.
Failure to meet IR 90-03 and LER Root cause attributed to design separation criteria for 90-08 deficiency. The NRC determined power cables to that the licensee's corrective actions Regulatory Guide 1.97 were acceptable (modification of the instrumentation instrument cabling).
Lack of redundancy in IR 89-09, IR Root cause attributed to an RHR service water 89-12, IR 91-03 inadequate modification design systems.
and LER 89-09 review. The NRC determined that the licensee's corrective actions were adequate (modification and engineering review).
A 3
Issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Failure to meet IR 90-10, IR Multiple root causes were technical speellications 9015, LER 9010 identified, including engineering for the emergency diesel interface with plant operations, generator (EDG) timeliness ofincorporating changes operational readiness into the Final Safety Analysis test.
Report (FSAR), adequacy of technical review and the accuracy of vendor supplied information.
The NRC determined that the licensee's corrective actions were adequate.
Reduced cooling water IR 91-13, IR Root cause attributed to an flow to the EDGs and 91-21, IR 91 19, inadequate safety evaluation. The station service air LER 91-12 NRC issued an SL III Violation for compressors due to high the licensee's failure to perform a sen' ice water system written safety evaluation as backpressure.
required by 10 CFR 50.59. The licensee developed a pmgram to strengthen its safety review process.
The NRC is reviewing the effectiveness of this program.
Removal c 'a technical IR 8917, LER Root cause attributed to inadequate specification 89 20 review of a procedural revision.
surveillance requirement The NRC determined that the from a procedure due to licensee's corrective actions were an inadequate review.
acceptable (detailed review of related procedures, procedural revision, and personnel instruction on review of technical specification procedures).
--.a 4
4 Issues Raised in 12/16/92 NECNP Letter
~
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Reactor scram resulting IR 91-07, LER Root cause attributed to component from a loss of off-site 91-05 failure (falled insulator in the 345-power (LOOP) that was kV switchyard). The NRC caused by the determined that the licensee's mechanical failure of a corrective actions wem adequate 345-kV switchyard bus (component repair, switchyard which was attributed to inspection).
a broken high voltage insulator stack.
Reactor scram caused IR 91-12, IR Root cause attributed to an by LOOP caused by an 91 13, IR 91 19, inadequate procedure (maintenance inadequate procedure IR 91-22, LER guideline). The NRC determined guideline.
91-09.
that the licensee's corrective actions were adequate to support plant restart.
Reactor scram due to IR 91-14,91-22, Root cause attributed to a lightning loss of a 345-kV and LER strike. Some equipment problems switchyard.
91-14.
were noted during the plant trip which were corrected-prior to plant startup.
Loss of all normal IR 87-16, LER Root cause attributed to a fault on power during shutdown 87-08 an off site transmission line and due to routing all off.
the routing of all off. site power site power sources sources through one breaker. The through one breaker.
NRC determined that the licensee's planned corrective actions (procedural revisions) were appropriate.
Degraded grid IR 92 09, IR Root cause attributed to setpoint undervoltage (UV) 93-04, LER drift. The NRC determined that relays found below 92-12 the licensee's planned corrective technical specification actions (reset relays and limits.
engineering evaluation) were appropriate.
o 4
5 lssues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Failed relay coil results IR 91 12, LER Root cause attributed to component in a primary' 91-10 failure (lack of an established containment isolation service life for norr. tally energized system actuation.
GE CR 120 relays). The NRC determined that the licensee's corrective actions (established a service life for this component) were appropriate.
Loss of "B" loop IR 9107, IR Root cause attributed to an shutdown cooling due to 91-19, IR 91 11, inadequate procedure for the pressure switch IR 92-01, LER system configuration. Final NRC actuation.
91-06 review of the licensee's corrective actions were adequate (procedurt revision, and engineering analysis is pending).
Failure to perform daily IR 8917, IR Root cause attributed to inadequate Instrument checks on 91-05, LER 89 23 procedural review. The NRC the low pressure coolant determined that the licensee's injection (LPCI) system corrective actions were good crosstle monitor.
(procedural revision and review).
Containment isolation IR 91-14, IR Root cause attributed to component valve failure to close 91-24, LER 91 15 failure (erosion of the screw in seat due to erosion / corrosion threads). The NRC determined and displacement of the that the licensee's corrective actions screw in seat.
were adequate (replaced valve seat and inspection of similar valves).
Turbine trip and reactor IR 88 08, IR Root cause attributed to component scram due to feedwater 88-10, LER 88-07 failure. The NRC determined that controller malfunction.
the licensee's corrective actions were adequate.
.h 6
Issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Appendix J Type B and IR 84-26, IR Root cause attributed to component C failure due to check 85-08, IR 88-10, failure. The NRC determined valve seat leakage.
IR 89 02, IR (NRC Inspection Report 50-271/89-90-15, IR
- 02) that the licensee's corrective 91-19, IR 92 09, actions prior the 1990 refueling IR 92-16, IR outage were ineffective. During the 92-18, LER 84-11, 1990 refueling outage, the licensee LER 89-07, replaced the valve's elastometric LER 90-12, LER seating material with stellite. This 92-10 valve (FDW 96A) passed its leak test during the 1992 refueling outage. The NRC determined that i
the licensee's frequent repair of this valve, Indicated earlier ineffective corrective action, but was not Indicative of a tolerance for a degraded condition.
Service water check IR 89-07, IR Root cause attributed to valves inoperable due to 91 19, LER 89-17 microbe-induced cormston (MIC).
corrosion of internal The NRC determined that the parts.
licensee's corrective actions were adequate (Increased Inspection frequency, and system piping upgrades).
Inadvertent scram and IR 92 06, IR Root cause attributed to personnel ECCS Initiation while 92-09, LER 92 014 error (poor coordination when restoring level restoring the vessel level reference transmitters to service.
legs to service). The NRC determined that the licensee's corrective actions were adequate (system restoration and personnel training).
.=
._n-7
.lssues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Missed diesel fire pump IR 91-07, IR Root cause attributed to an fuel oil surveillance due 91-11, IR 91-14, inadequate procedure. 'Ihe NRC to inadequate
- 91-19 LER 9103 issued a non-cited violation to the procedure, licensee, and concluded that the -
licensee's corrective actions were adequate (oil sampled, procedure revised, review of all ehemistry department procedures, and training enhancements).
Reactor vessel inventory IR 89 02, LER Root cause attributed to personnel decrease due to 89-13 error. The NRC determined that personnel error.
the licensee's corncthe actions were adequate (procedural revisions and administrative controls).
Plant service water IR 88-03, IR Root cause attributed to an emuent stream not 88-06, IR 8814, inadequate procedure. The NRC monitored due to an LER 88-01 determined that the licensee's inadequate procedure.
corrective actions were not properly implemented and issued a Notice of Violation (Severity Level IV) for this event.
Missed emuent sample IR 88-14, IR 88 20 Root cause attributed to inadequate due to inadequate LER 8814 Implementation of corrective corrective action in LER actions for LER 88 01. The NRC 88-01.
Issued a Notice of Violation to the licensee for falling to meet this technical specification requirement.
The licensee took additional corrective actions (procedural and log sheet revisions, and personnel training) which NRC determined were adequate.
i 8
-issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION Missed RHR valve IR 89-22, LER Root cause attributed to an leakage surveillance due 89 24 Inadequate procedural review. The to incomplete procedure licensee's investigation determined review.
that this valve was not required to be in the IST Program and was subsequently removed from the program. The NRC determined that the licensee's corrective actions were appropriate.
Priman containment IR 90-15, IR Root cause attributed to a weakness isolation system (PCIS) 9018, LER 90-18 in the testing procedure. The NRC spurious actuation due determined that the licensee's to an inadequate corrective actions were appropriate procedure.
(procedural revisions).
Improper inservice flow IR 9216, LER Root cause attributed to personnel testing of the control 92 16 error. The NRC determined that room chilled water the licensee's corrective actions pump due to ASME were adequate (procedure stvisions, Code misinterpretation IST Program review, personnel and subsequent missed training).
quarterly test due to incorrectly following the test procedure.
"A" emergency diesel IR 91-19 Root cause attributed to component generator fuel oil failure. The NRC determined that transfer pump the licensee's post maintenance operability.
testing was adequate to ensure operability.
"B" emergency diesel IR 91-19 Root cause attributed to component generator failure to failure (wear of the fuel oil safety start.
valve). 'Ihe NRC dete mined that the licensee's Investigation and repair efforts were thoruugh.
. _.c.g 4
9 Issues Raised in 12/16/92 NECNP Letter
SUMMARY
REFERENCES ROOT CAUSE AND RESOLUTION
'B' EDG failure to start IR 92 06 Root cause(s) attributed to during ECCS testing.
component malfunction (s). The first failure was due to incomplete resetting of diesel governor shutdown plunger which the licensee attributed at first to the advanced age of the governor. A second failure occurred when the diesel generator output breaker failed to shut due to binding on an auto start relay. The NRC determined that the licensee's repairs and investigations were adequate.
i 25i n
ENCLOSURE 2 VERMONT YANKEE LERs 1987 THROUGH 1992 Of the 115 ERs issued in the 6-year period from 1987 through 1992, a majority (36%)
involved personnel errors - not an uncommon cause given the period of time or complexity of operations and testing. Deficient procedures account for more than 20% of the events involved, and this is an issue for human performance and administrative controls (and not equipment degradation). Smaller fractions for design deficiencies, external, other and unknown causes account for less than 20% of the events.
CAUSE CODES SUBTOTAIS FUNCTIONAL AREA A
B
_C D
E X
(BY AREA)
Operations 17 1
1 6
17 7
49 Radiological Controls 7
0 0
1 0
0 8
Maintenance and Surveillance 16 0
1 11 11 4
Security 0
Engineering and Technical Support 1
8 0
5 1
0 15 Safety Assessment /
Ouality Verification 0
Subtotals 41 9
2 23 29 11 115 Notes:
- 1. ERs issued from 1987 through 1992
- 2. Cause codes identified are be. sed upon NRC staff evaluations and inspections of the events, and may in certain instances differ from those specified in the ER.
- 3. Cause codes are:
A.
Personnel Error B.
Design C.
External or Unknown D.
Procedure Inadequacy E.
Component Failure X.
Other
- 4. ERs in the functional area of security do not include safeguards reports under 10 CFR Part 73.
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9' 2
The balance (29 events or approximately 259) involve component failures. In evaluating the potential de~ gradation of safety equipment due to deficiencies in test or rnaintenance practices, reportable events caused by component failures are of interest. A breakdown of these component failures over the six-year period shows that relatively few of the LERs caused by component failure were attributed to deficiencies in maintenance and surveillance test programs. Such deficiencies would be otherwise indicative of weakness in predictive and preventive methods. A chronological breakdown by SALP period is provided below and indicates that there is n6 increasing trend over the period in question.
COMPONENT FAILURES SALP REPORT DUE TO MAINTENANCE PERIOD NUMBER TOTAL OR TEST 1/87 - 6/88 87-99 9
3 7/88 - 9/89 88-99 7
3 10/89 - 3/91 89-99 5
0 3/91 - 8/92 91-99 7
4 8/92 - 12/92 Current 1
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